added nuc data config for task 4

fusion-energy · Oct 9, 2024 · 295ca24 · 295ca24
1 parent c5b38e5
commit 295ca24
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Showing 3 changed files with 58 additions and 5 deletions.
diff --git a/tasks/task_08_CSG_mesh_tally/1_example_2d_regular_mesh_tallies.ipynb b/tasks/task_08_CSG_mesh_tally/1_example_2d_regular_mesh_tallies.ipynb
@@ -27,6 +27,26 @@
     "IFrame(src=\"https://www.youtube.com/embed/KYIsDjip1nQ\", width=560, height=340)"
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import packages needed and configure the OpenMC nuclear data path"
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
+    "import matplotlib.pyplot as plt\n",
+    "import openmc\n",
+    "# Setting the cross section path to the correct location in the docker image.\n",
+    "# If you are running this outside the docker image you will have to change this path to your local cross section path.\n",
+    "openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'"
+   ]
+  },
   {
    "cell_type": "markdown",
    "metadata": {},
@@ -42,9 +62,6 @@
    "metadata": {},
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "import matplotlib.pyplot as plt\n",
-    "\n",
     "# MATERIALS\n",
     "\n",
     "# creates two materials, one is a neutron multiplier (lead) and the other a tritium breeder (lithium)\n",

diff --git a/tasks/task_08_CSG_mesh_tally/2_example_3d_regular_mesh_tallies.ipynb b/tasks/task_08_CSG_mesh_tally/2_example_3d_regular_mesh_tallies.ipynb
@@ -20,6 +20,25 @@
     "This first code block defines the model geometry, materials and neutron source."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import OpenMC and configure the nuclear data path"
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
+    "import openmc\n",
+    "# Setting the cross section path to the correct location in the docker image.\n",
+    "# If you are running this outside the docker image you will have to change this path to your local cross section path.\n",
+    "openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,
@@ -28,8 +47,6 @@
    },
    "outputs": [],
    "source": [
-    "import openmc\n",
-    "\n",
     "# MATERIALS\n",
     "\n",
     "breeder_material = openmc.Material()   # Pb84.2Li15.8\n",

diff --git a/tasks/task_08_CSG_mesh_tally/3_example_cylinder_mesh_tallies.ipynb b/tasks/task_08_CSG_mesh_tally/3_example_cylinder_mesh_tallies.ipynb
@@ -18,6 +18,25 @@
     "This first code block defines the model geometry and materials."
    ]
   },
+  {
+   "cell_type": "markdown",
+   "metadata": {},
+   "source": [
+    "First import OpenMC and configure the nuclear data path"
+   ]
+  },
+  {
+   "cell_type": "code",
+   "execution_count": null,
+   "metadata": {},
+   "outputs": [],
+   "source": [
+    "import openmc\n",
+    "# Setting the cross section path to the correct location in the docker image.\n",
+    "# If you are running this outside the docker image you will have to change this path to your local cross section path.\n",
+    "openmc.config['cross_sections'] = '/nuclear_data/cross_sections.xml'"
+   ]
+  },
   {
    "cell_type": "code",
    "execution_count": null,